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Merge pull request #860 from wbinventor/hotfix-scatt-mat-nuclides

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Bug fix for consistent scattering matrix with nuclides
This commit is contained in:
Adam Nelson 2017-04-12 19:23:27 -04:00 • committed by GitHub
commit bc4683be1c
2 changed files with 21 additions and 9 deletions
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28
openmc/mgxs/mgxs.py
View file

@ -3917,6 +3917,9 @@ class ScatterMatrixXS(MatrixMGXS):
scatter_p0 = self.tallies['{}-0'.format(self.rxn_type)]
scatter_p1 = self.tallies['{}-1'.format(self.rxn_type)]
energy_filter = scatter_p0.find_filter(openmc.EnergyFilter)
# Transform scatter-p1 tally into an energyin/out matrix
# to match scattering matrix shape for tally arithmetic
energy_filter = copy.deepcopy(energy_filter)
scatter_p1 = scatter_p1.diagonalize_filter(energy_filter)
self._rxn_rate_tally = scatter_p0 - scatter_p1
@ -3956,10 +3959,6 @@ class ScatterMatrixXS(MatrixMGXS):
self._xs_tally = MGXS.xs_tally.fget(self)
else:
# Compute groupwise scattering cross section
self._xs_tally = self.tallies['scatter'] / \
self.tallies['flux (tracklength)']
# Compute scattering probability matrix
energyout_bins = [self.energy_groups.get_group_bounds(i)
for i in range(self.num_groups, 0, -1)]
@ -3990,8 +3989,13 @@ class ScatterMatrixXS(MatrixMGXS):
# Remove the AggregateFilter summed across mu bins
norm._filters = norm._filters[:2]
# Multiply by the group-to-group probability matrix
self._xs_tally *= (self.tallies[tally_key] / norm)
# Compute groupwise scattering cross section
self._xs_tally = self.tallies['scatter'] * \
self.tallies[tally_key] / norm / \
self.tallies['flux (tracklength)']
# Override the nuclides for tally arithmetic
self._xs_tally.nuclides = self.tallies['scatter'].nuclides
# Multiply by the multiplicity matrix
if self.nu:
@ -4001,14 +4005,22 @@ class ScatterMatrixXS(MatrixMGXS):
# If using P0 correction subtract scatter-1 from the diagonal
if self.correction == 'P0' and self.legendre_order == 0:
flux = self.tallies['flux (analog)']
scatter_p1 = self.tallies['{}-1'.format(self.rxn_type)]
flux = self.tallies['flux (analog)']
# Transform scatter-p1 tally into an energyin/out matrix
# to match scattering matrix shape for tally arithmetic
energy_filter = flux.find_filter(openmc.EnergyFilter)
energy_filter = copy.deepcopy(energy_filter)
scatter_p1 = scatter_p1.diagonalize_filter(energy_filter)
self._xs_tally -= (scatter_p1 / flux)
# Compute the trasnport correction term
correction = scatter_p1 / flux
# Override the nuclides for tally arithmetic
correction.nuclides = scatter_p1.nuclides
self._xs_tally -= correction
self._compute_xs()
return self._xs_tally

2
tests/test_mgxs_library_nuclides/results_true.dat
View file

@ -1 +1 @@
107576b21fa8ed72f71ac65866e4ad1100cc62df527f6f4e6bb8a6318f3694614b6ba1c72634b2e97474a75ea1b3112dcb9b4b36e58f0417e5b04aee5d3c7570
ea1c7567cb5399ab26297b5df0c1c21b52863303746d04f1698760d88b1aafb625d5743969ad30b194c353d106ed4287a8bc755ea6404c2f5dafb2ea930444a7